Molecular dynamics simulation of beryllium oxide irradiated by deuterium ions: sputtering and reflection

The sputtering and reflection properties of wurtzite beryllium oxide (BeO) subjected to deuterium (D) ions bombardment at 300 K with ion energy between 10 eV and 200 eV is studied by classical molecular dynamics. Cumulative irradiations of wurtzite BeO show a D concentration threshold above which an 'unphysical dramatic' sputtering is observed. From the cumulative irradiations, simulation cells with different D concentrations are used to run non-cumulative irradiations at different concentrations. Using a D concentration close to the experimentally determined saturation concentration (0.12 atomic fraction), the simulations are able to reproduce accurately the experimental sputtering yield of BeO materials. The processes driving the sputtering of beryllium (Be) and oxygen (O) atoms as molecules are subsequently determined. At low irradiation energy, between 10 eV and 80 eV, swift chemical sputtering (SCS) is dominant and produces mostly ODz molecules. At high energy, the sputtered molecules are mostly BexOy molecules (mainly BeO dimer). Four different processes are associated to the formation of such molecules: the physical sputtering of BeO dimer, the delayed SCS not involving D ions and the detachment-induced sputtering. The physical sputtering of BeO dimer can be delayed if the sputtering event implies two interactions with the incoming ion (first interaction in its way in the material, the other in its way out if it is backscattered). The detachment-induced sputtering is a characteristic feature of the 'dramatic' sputtering and is mainly observed when the concentration of D is close to the threshold leading to this sputtering regime.Peer reviewe

Hydrogen supersaturated layers in H/D plasma-loaded tungsten: A global model based on thermodynamics, kinetics and density functional theory data

International audienceIn this work, we combine Density Functional Theory data with a Thermodynamic and a kinetic model to determine the total concentration of hydrogen implanted in the sub-surface of tungsten exposed to a hydrogen flux. The sub-surface hydrogen concentration is calculated given a flux of hydrogen, a temperature of implantation, and the energy of the incoming hydrogen ions as independent variables. This global model is built step by step; an equilibrium between atomic hydrogen within bulk tungsten and a molecular hydrogen gas phase is first considered, and the calculated solubility is compared with experimental results. Subsequently, a kinetic model is used to determine the chemical potential for hydrogen in the sub-surface of tungsten. Combining both these models, two regimes are established in which hydrogen is preferentially trapped at either interstitial sites or in vacancies. We deduce from our model that the existence of these two regimes is driven by the temperature of the implanted tungsten sample; above a threshold or transition temperature is the interstitial regime, below is the vacancy regime in which super-saturated layers form within tenths of angstrom below the surface. A simple analytical expression is derived for the coexistence of the two regimes depending on the implantation temperature, the incident energy and the flux of the hydrogen ions which we use to plot the corresponding phase diagram

Identification of BeO and BeOxDy in melted zones of the JET Be limiter tiles: Raman study using comparison with laboratory samples

Beryllium oxide (BeO) and deuteroxide (BeOxDy) have been found on the melted zone of a beryllium tile extracted from the upper dump plate of JET-ILW (2011-2012 campaign). Results have been obtained using Raman microscopy, which is sensitive to both the chemical bond and crystal structure, with a micrometric lateral resolution. BeO is found with a wurtzite crystal structure. BeOxDy is found as three different types which are not the beta-phase but behaves as molecular species like Be(OD)(2), O(Be-D)(2) and DBeOD. The presence of a small amount of trapped D2O is also suspected. Our results therefore strongly suggest that D trapping occurs after melting through the formation of deuteroxides. The temperature increase favors the formation of crystal BeO which favors deuterium trapping through OD bonding.EURATOM 63305

Tritium retention in W plasma-facing materials : Impact of the material structure and helium irradiation

This article has an erratum: DOI 10.1016/j.nme.2020.100729Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of W samples: first, different types of W materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure W samples submitted to various annealing and manufacturing procedures, along with monocrystalline W for reference. Then, He and He-D irradiated W samples were studied to investigate the impact on He-damages such as nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 degrees C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption as a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of "as received" industrial W (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in "as received W" compared to annealed and polish W, and desorbs only at 800 degrees C, meaning ideal W material studies may underestimate T inventory for tokamak relevant conditions.Peer reviewe

Impact of W Events and Dust on JET-ILW Operation

ABSTRACT The occurrence of transient impurity events (TIE) leading to intense radiation spikes in JET plasma discharges has been studied since the installation of the ITER-like Wall (ILW). To generate the observed average increase in radiated power of 1.5MW, a 100µm-radius sphere of solid W dust would be required. The drop in plasma energy caused by W-TIEs is fully recovered in 90% of all cases, only 1% inducing a longer term loss in plasma energy which sometimes leads to the shutdown of plasma operation. TIEs are correlated with disruptions and with measurements of the dust mobilized by disruptions using the high resolution Thomson scattering (HRTS) diagnostic. The dust characteristics giving rise to TIEs have been studied using the dust transport code DTOKS and the 1D impurity transport code STRAH

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate

Dynamic modelling of local fuel inventory and desorption in the whole tokamak vacuum vessel for auto-consistent plasma-wall interaction simulations

An extension of the SolEdge2D-EIRENE code package, named D-WEE, has been developed to add the dynamics of thermal desorption of hydrogen isotopes from the surface of plasma facing materials. To achieve this purpose, DWEE models hydrogen isotopes implantation, transport and retention in those materials. Before launching autoconsistent simulation (with feedback of D-WEE on SolEdge2D-EIRENE), D-WEE has to be initialised to ensure a realistic wall behaviour in terms of dynamics (pumping or fuelling areas) and fuel content. A methodology based on modelling is introduced to perform such initialisation. A synthetic plasma pulse is built from consecutive SolEdge2D-EIRENE simulations. This synthetic pulse is used as a plasma background for the D-WEE module. A sequence of plasma pulses is simulated with D-WEE to model a tokamak operation. This simulation enables to extract at a desired time during a pulse the local fuel inventory and the local desorption flux density which could be used as initial condition for coupled plasma-wall simulations. To assess the relevance of the dynamic retention behaviour obtained in the simulation, a confrontation to post-pulse experimental pressure measurement is performed. Such confrontation reveals a qualitative agreement between the temporal pressure drop obtained in the simulation and the one observed experimentally. The simulated dynamic retention during the consecutive pulses is also studied

First mirror test in JET for ITER: Complete overview after three ILW campaigns

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015–2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011–2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20–80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5×10$^{17}$ cm$^{-2}$, W is up to 4.3×10$^{17}$ cm$^{-2}$, while the content of Ni is the greatest in the outer divertor: 3.8×10$^{17}$ cm$^{-2}$. Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1×10$^{16}$ cm$^{-2}$. The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s$^{-1}$) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred

First mirror test in JET for ITER : complete overview after three ILW campaigns

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011-2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 x 10(17) cm(-2), W is up to 4.3 x 10(17) cm(-2), while the content of Ni is the greatest in the outer divertor: 3.8 x 10(17) cm(-2). Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10(16) cm(-2). The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s(-1)) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred

Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015–2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and β-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013–2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign